ML20071L826
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| ML20071L826 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 09/16/1982 |
| From: | Shamla W NORTHERN STATES POWER CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 8209240113 | |
| Download: ML20071L826 (1) | |
Similar Documents at Monticello | |
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Category:REPORTABLE OCCURRENCE REPORT (SEE ALSO AO LER)
MONTHYEARML20202F9161998-01-29029 January 1998
[Table view]Special Rept:On 980128,two of Three Fire Protection Pumps Were Removed from Svc as Result of Sys Configuration Necessary to Support Planned Maintenance.Fire Pumps Were Returned to Svc on 980128 ML20138F5801997-04-29029 April 1997 Special Rept:On 970402,declared Reactor Bldg Wide Range Gas Monitor Inoperable.Caused by Ruptured Pump Diaphragm. Initiated Work Order,Installed Replacement Pump & Declared Pump Operable ML20137Y2171997-04-15015 April 1997 Special Rept:On 961220,util Rail Car Was Released to Burlington Northern Railroad from Monticello Nuclear Plant. Rail Car Was Delayed in Chicago as Result of Problems W/Bill of Lading on Computer Sys.Rail Car Was Delivered on 970131 ML20101E8321996-03-15015 March 1996 Special Rept:On 960314,two of Three Fire Pumps Removed from Svc for Planned Mod Work.Returned Both Fire Pumps to Svc Same Day ML20094E4371995-10-31031 October 1995 Special Rept:On 951010,sys Engineer & Technicians Reintiated Work on B Channel of Reactor Bldg Vent Radiation Monitor, After Providing Proper Notification to Main Control Room. Caused by Incorrect Channels Removed from Svc ML20082R6601995-04-19019 April 1995 Special Rept:On 950404,discovered That Manual Isolation Valves for 'A' Channel in Closed Position.Cause of Failure to Correctly Position Valves Will Be Assessed.Valves Placed in Required Position & Demonstrated Operability ML20082E5661995-04-0606 April 1995 Special Rept:On 950326,pump Discharge Relief Valve for Electric Screen Wash/Fire Pump Became Inoperable.Valve Had to Be Secured to Permit Installation of an Isolation Device. Pump Returned to Operational Status After Device Installed ML20080G4731995-02-0101 February 1995 Special Rept:On 950119,two of Three Fire Protection Pumps Out of Svc for Period of Approx 15 H.Action Taken for Planned Preventative Maint on Two Check Valves & Drain Valve in Fire Suppression Sys.Roving Fire Watch Established ML20067D1351994-02-23023 February 1994 Special Rept:On 940126,fire Door 105 Declared Inoperable Due to Inability to Close Properly.Caused by Higher than Normal Differential Pressure Between Plant Administration Bldg & Turbine Bldg.Cause of Failure Corrected ML20117A5801992-11-24024 November 1992 Special Rept:On 921020,determined That RHR Sample Supply Loops a & B Excess Flow Check Valves Not Included in ASME Section XI Program.Caused by Lack of Proven Test Procedure. Valves Incorporated Into Third 10-yr ASME Program ML20082G9261991-08-15015 August 1991 Special Rept:On 910719,potential for Inoperable Penetration Fire Barrier Identified.Caused by Use of Less Conservative, But Technically Acceptable,Design Alternatives in Plant Const.Vents Will Be Rerouted ML20244E5941989-06-16016 June 1989 Special Rept:On 890603,two Out of Three Diesel Fire Pumps Were Inoperable for Less than 12 H.Caused by Diesel Fire Pump Day Tank Outlet Valve Being Closed Instead of Fill Valve During Performance Test 1158.Test Revised ML20081G9561983-10-25025 October 1983 Ro:On 831024,trip Coil for Reactor Recirculation Motor Generator Set 11 Drive Breaker Failed to Trip Automatically. Investigation Ongoing.Trip Coil Replaced ML20066E0131982-11-0101 November 1982 Ro:On 821031,thru Wall Defect Found on C 12-inch Recirculation safe-end to Pipe Weld Joint.Resolution Under Investigation ML20064E5141982-10-21021 October 1982 Ro:On 821020,thru Wall Defect Found on E 12-inch Recirculation Sys Safe End to Pipe Weld Joint.Indications Will Be Documented as Revision to RO 82-16 Reported on 820928.Resolution Being Investigated ML20064E4761982-10-11011 October 1982 Ro:On 821009,linear Indications Confirmed to Exist on Three Addl Welds in Recirculation Sys.Indications Will Be Documented as Revision to RO 82-16 Reported on 820928. Resolution Being Investigated ML20071N4361982-09-28028 September 1982 Ro:On 820928,crack Indication in End Cap of a Recirculation Riser Confirmed by Radiography.Indications Initially Detected by Ultrasonic Exam During Normal Inservice Insp. Remedial Measures Under Investigation ML20071L8261982-09-16016 September 1982 Ro:On 820915,leak Discovered on Primary Containment Suppression Chamber Nitrogen Control Sys Inboard Isolation Valve (AO-2378).Leakage Less than Tech Spec Allowable Leakage ML20053E8071982-06-0202 June 1982 RO Iii:On 810224,failure of Discharge Valve on Instrument & Svc Air Compressor Resulted in Loss of Instrument Air Sys Pressure Causing Plant Scram.Failed Check Valve Replaced W/Similar Unit ML20058L3431978-07-31031 July 1978 Ro:On 780728,control Rod Drive 30-47 Delayed Approx 1.4 Before Scramming.Buna N Disk in Plunger Broken in Pieces.Plungers on 242 Scram Pilot Valves & Two Backup Scram Pilot Valves Removed & Replaced ML20090L8311978-06-0202 June 1978 Ro:On 780505,steam Leak Noted in RCIC Inlet Steam Line & Drain Line to Condenser.Caused by Pinhole Failure on Weld on 300 Lb Socket Weld.Hole Temporarily Patched ML20091A9941978-03-10010 March 1978 Ro:On 780309,smear Survey of Chem Nuclear Sys Inc Model 4-45 Shipping Cask Disclosed Max Surface Contamination of 25,900 Disintegrations Per Min Per 100 Square Centimeters.Cask Will Be Decontaminated ML20127G6171978-02-0202 February 1978 Advises That on 780202,during Surveillance Test,One Suppression Chamber to Drywell Vacuum Breaker Failed to Reclose Following Exercise.Plant Shutdown Initiated ML20086D5141978-01-0606 January 1978 Telecopy Ro:On 780105,discovered That One of Two Nuclear Engineering Co Model B3-1 Shipping Casks Provided w/1-inch Diameter Lid Bolts Rather than 1-1/4 Inch Size Specified in Certificate of Compliance 6058 ML20086D5201977-12-14014 December 1977 Telecopy Ro:Heating Steam Coil Leak in Reactor Bldg Ventilation Supply Unit V-AH-4A Resulted in Freezing of Condensate at Inlet to Unit & Subsequent Inoperability of Associated Secondary Containment Isolation Dampers ML20086D5371977-10-14014 October 1977 Telecopy Ro:Insp of Internal Torus Catwalk Support Structure Revealed That Catwalk Mitered Sections Not All Attached to Horizontal Catwalk Support Plates in Some Manner.Attachment Locations Will Be Upgraded ML20086D5501977-10-13013 October 1977 Telecopy Ro:On 771012,local Leak Rate Testing of MSIV AO-2-80A & Nitrogen Instrument Air Sys Isolation Valve CV-7436 Indicated That Leakage Exceeded Tech Spec Acceptance Criteria.Cause Under Investigation ML20086D5541977-10-0505 October 1977 Telecopy Ro:On 771004,local Leak Rate Testing of HPCI Sys Discharge Isolation Check Valve HPCI-9 Indicated That Leakage Exceeded Tech Spec Acceptance Criteria.Cause Under Investigation ML20086D5571977-09-29029 September 1977 Telecopy Ro:On 770929,local Leak Rate Testing of Core Spray Inboard Isolation Check Valve AO-14-13A Indicated That Leakage Exceeded Tech Spec Acceptance Criteria.Cause Under Investigation ML20086D5611977-09-14014 September 1977 Telecopy Ro:On 770913,local Leak Rate Testing of Core Spray Inboard Isolation Check Valve AO 14-13B Indicated That Leakage Exceeded Tech Spec Acceptance Criteria.Related Event on 770914 ML20086D5671977-09-12012 September 1977 Telecopy Ro:Main Steam Drain Isolation Valve MD 2373 & Main Steam Outboard Isolation Valve AO 2-86A Exceeded Tech Spec Acceptance Criteria for Local Leak Rate Tests ML20086D6421977-07-11011 July 1977 Telecopy Ro:On 770709,determined That Recombiner Sys a Offgas Flow Control Valve PCV 7489A Could Be Opened W/ Controlling Solenoid Valve Deenergized.Valve Held from Svc Pending Investigation ML20058K6111977-03-0909 March 1977 Ro:On 770223,discovered Withdrawal of in-sequence Control Rod Resulted in Period Less than 5-s & IRM Scram ML20086D6911977-03-0202 March 1977 Telecopy Ro:On 770301,torus Water Vol Found to Be Slightly Below Min Vol Established by Tech Specs.Water Vol Restored to Normal Operating Level ML20086D7031977-02-24024 February 1977 Telecopy Ro:On 770223,while Withdrawing in-sequence Control Rod to Bring Reactor Critical,Reactor Period of Less than 5 Obtained ML20086D7881976-09-10010 September 1976 Telecopy Ro:On 760909,discovered That Torus Water Vol Several Hundred Cubic Ft Below 68,000 Cubic Ft Min Required by Tech Specs.Caused by Failure to Correct for Vol Vs Level Correlation ML20056B8541976-05-21021 May 1976 Informs of 760520 Event Involving TIP Bell Valve Which Failed to Properly Close During Operation of TIP Sys.Valve Cleaned,Relubricated & Closure Spring Tension Increased ML20086D8231976-05-0505 May 1976 Telecopy Ro:On 760504,during Routine Surveillance Test, Primary Containment Oxygen Concentration Found to Be in Excess of Tech Spec Limit.Caused by Leakage from Drywell Instrument Air Sys Into Containment Nitrogen Supply Line ML20127L2611975-09-22022 September 1975 Informs of 750921 Incident Re Countrate Decreasing Below 3 Counts Per While Performing Refueling Core Alterations. Countrate Increased & Refueling Operations Recommenced. Further Investigation Continuing ML20127L2751975-09-22022 September 1975 Informs of 750920 Incident Re Drywell Equipment Drain Sump Isolation Valves AO 2561A & B Exceeding TS Acceptance Criteria.Further Investigation Continuing ML20127L2551975-09-19019 September 1975 Informs That on 750919,during Local Leak Rate Testing of MSIVs 2-86A & 2-88A,combined Leakage of Valves Exceeded Acceptance Criteria.Further Investigation Continuing ML20127L2391975-09-15015 September 1975 Informs of 750914 Incident Re Discovery That Hydraulic Shock Suppressor Located Inside Primary Containment on Loop B HPCI Line Inoperable Due to Loss of Oil.Reactor in Cold Shutdown for Refueling.Further Investigation Underway ML20127L1741975-08-21021 August 1975 Discusses 750818 Incident Re Broken Seal Cooling Line at Threaded Connection on Reactor Water Cleanup Pump 12.Pump Immediately Isolated & Line Repaired.Investigation Into Cause of Failure Continuing ML20127L2651975-07-28028 July 1975 Ro:On 750727,operator Motor Failed & Stalled Due to Stalled Rotor Current Overheating Windows.Investigation Pending ML20127L2521975-07-14014 July 1975 Ro:On 750713,small Leak Developed on Local Pressure Gauge Monitoring Reactor Pressure.Root Valve for Pressure Gauge Closed,Allowing Ref Leg to Refill & Water Level Indication to Return to Normal ML20058L1761975-07-11011 July 1975 Ro:On 750624 & 25,10 Initial Core Fuel Assemblies Which Had Been Identified as Leakers & Removed from Core at End of Cycles 2 & 3 Inspected Using Underwater Television ML20058K6751975-07-0303 July 1975 Notifies That on 750703,during Routine Surveillance Testing, Reactor High Pressure Scram Switches Found Inoperable ML20127B7371975-05-27027 May 1975 Ro:On 750525 Core Spray Sys Motor Operated Valve Failed to Reopen by Means of Motorized Operator.Valve Was Returned to Normal Open Position by Local Handwheel ML20058K6951975-05-0505 May 1975 Ro:On 750505,discovered That Control Circuit for Valve Steam Supply to RCIC Turbine Was Deenergized by Motor Control Unit Undervoltage Relay Coil Opening.Coil Replaced & Remote Operability of Valve Demonstrated ML20127L1471975-05-0101 May 1975 RO 75-12:on 750429,surveillance Test for Exercising Unit 2 Containment Vacuum Breaker Isolation Valve CV-31628 Showed Valve to Be Inoperable 1998-01-29 Category:TEXT-SAFETY REPORT MONTHYEARML20217D1261999-09-30030 September 1999
[Table view]Monthly Operating Rept for Sept 1999 for Monticello Nuclear Generating Plant.With ML20216E7031999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Monticello Nuclear Generating Plant.With ML20210Q0521999-08-0404 August 1999 Safety Evaluation Approving Relief Request 10 to License DPR-22 Per 10CFR50.55a(g)(6)(i).Inservice Exam for Relief Request 10,Parts A,B,C,D & E Impractical & Reasonable Assurance of Structural Integrity Provided ML20210Q6611999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Monticello Nuclear Generating Plant.With ML20209F7901999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Monticello Nuclear Generating Plant.With ML20195H0351999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Monticello Nuclear Generatintg Plant.With ML20206N1721999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Monticello Nuclear Generating Plant.With ML20205N0861999-04-12012 April 1999 Safety Evaluation Re Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20205P5701999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Monticello Nuclear Generating Plant.With ML20204H4951999-03-18018 March 1999 SER Concluding That Licensee Established Acceptable Program to Verify Periodically design-basis Capability of safety-related MOVs at Monticello.Therefore Staff Concludes Licensee Adequately Addressed Action Required in GL 96-05 ML20205G7391999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Monticello Nuclear Generating Plant.With ML20202F7901999-01-25025 January 1999 1999 Four Year Simulator Certification Rept for MNGP Simulation Facility ML20199E4871999-01-0606 January 1999 SER Accepting Licensee 951116,960214 & 0524 Responses to NRC Bulletin 95-002, Unexpected Clogging of Residual Heat Removal Pump Strainer While Operating in Suppression Pool Cooling Mode ML20199F6211998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for Mngp.With ML20205H0561998-12-31031 December 1998 Northern States Power Co 1998 Annual Rept. with ML20198P0691998-12-28028 December 1998 Safety Evaluation Concluding That NSP Proposed Alternative to Paragraph III-3411 of App III to 1986 Edition of Section XI of ASME Code Provides Acceptable Level of Quality & Safety.Alternative Authorized ML20198D0751998-12-10010 December 1998 Safety Evaluation Supporting NSP Proposed Change to EOPs to Use 2/3 Core Height as Potential Entry Condition Into Containment Flooding ML20198B2531998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Monticello Nuclear Generating Plant.With ML20195E3691998-11-12012 November 1998 Safety Evaluation Concluding That Licensee USI A-46 Implementation Has Met Purpose & Intent of Criteria in GIP-2 & Staff Sser 2 for Resolution of USI A-46 ML20195D2381998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Monticello Nuclear Generating Plant.With ML20198J4451998-10-22022 October 1998 Rev 2 to SIR-97-003, Review of Test Results of Two Surveillance Capsules & Recommendations for Matls Properties & Pressure-Temp Curves to Be Used for Monticello Rpv ML20154L3471998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Monticello Nuclear Generating Plant.With ML20153F0511998-09-21021 September 1998 Rev 2 to MNGP Colr,Cycle 19 ML20153E9361998-09-0808 September 1998 Rev 1 to MNGP Colr,Cycle 19 ML20153B0861998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for Monticello Nuclear Generating Plant.With ML20237B8461998-07-31031 July 1998 Monthly Operating Rept for July 1998 for Monticello Nuclear Generating Plant ML20236W5041998-07-21021 July 1998 ISI Exam Summary Rept - Refueling Outage 19 ML20236R1941998-06-30030 June 1998 Monthly Operating Rept for June 1998 for Monticello Nuclear Generating Plant ML20249A5861998-05-31031 May 1998 Monthly Operating Rept for May 1998 for Monticello Nuclear Generating Plant ML20247K3971998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for Monticello Nuclear Generating Plant ML20217D8731998-04-13013 April 1998 Rev 0 to MNGP Colr,Cycle 19 ML20217F6431998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for Monticello Nuclear Generating Plant ML20216D1041998-03-0505 March 1998 Rev 21 to Operational QA Plan ML20216H6481998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for Monticello Nuclear Generating Plant ML20203G1431998-02-10010 February 1998 Rev 2 to Inservice Insp Exam Plan,Third Interval,920601- 020531 ML20203B2821998-01-31031 January 1998 Monthly Operating Rept for Jan 1998 for Monticello Nuclear Generating Station ML20202F9161998-01-29029 January 1998 Special Rept:On 980128,two of Three Fire Protection Pumps Were Removed from Svc as Result of Sys Configuration Necessary to Support Planned Maintenance.Fire Pumps Were Returned to Svc on 980128 ML20216D2071997-12-31031 December 1997 1997 Annual Rept for Northern States Power Co ML20197J8131997-12-31031 December 1997 Revised Evacuation Time Estimates for Plume Exposure Pathway Emergency Planning Zone at Monticello Nuclear Power Plant. W/One Oversize Drawing ML20198P2201997-12-31031 December 1997 Monthly Operating Rept for Dec 1997 for Monticello Nuclear Generating Plant ML20203J7131997-11-30030 November 1997 Monthly Operating Rept for Nov 1997 for Monticello Nuclear Generating Plant ML20199G7051997-11-19019 November 1997 Safety Evaluation Authorizing Relief Request 8 of Third 10 Yr Inservice Insp Interval ML20199H8181997-10-31031 October 1997 Monthly Operating Rept for Oct 1997 for Monticello Nuclear Generating Plant ML20217K2081997-09-30030 September 1997 Monthly Operating Rept for Sept 1997 for Monticello Nuclear Generating Plant ML20216H7771997-08-31031 August 1997 Monthly Operating Rept for Aug 1997 for Monticello Nuclear Generating Plant ML20217K2741997-07-31031 July 1997 Monthly Operating Rept for Jul 1997 for Monticello Nuclear Generating Plant ML20196H1081997-07-0808 July 1997 Rev 20 to Operational QA Plan ML20141B9271997-06-30030 June 1997 LOCA Containment Analyses for Use in Evaluation of NPSH for RHR & Core Spray Pumps ML20149E2921997-06-30030 June 1997 Monthly Operating Rept for June 1997 for MNGP ML20148S6341997-06-23023 June 1997 NPSH - Rept of Sulzer Bingham Pump 1999-09-30 |
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bm/S Mp NORTHERN STATES POWEM COMPANY Monticello Nuclear Generating Plant P.O. Box 600 Monticello, Minnesota 55362 September 16, 1982 Mr. J. G. Xeppler, Director Region 111 Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137
Dear Hr. Keppler,
MONTICELTA NUCLEAR GEhT_ RATING PLANT Docket No. 50-263 License No. DPR-22 On September 15, 1982 the following valve was found to leak in excess of the acceptance criteria for local leak tests specified in Section 4.7.A.2.f of tha Technical Specifications:
Primary Containment Suppression Chamber NitroRen Control System Inboard Isolation Valve, A0-2378.
It is noted that the Primary Containment Nitrogen Control System Outboard Isolation Valve had a leakage much less than technical specification allowable leakage.
Yours very truly,
- 7 Ll!.
^
W. A. Shamla 2 Plant Manager 8209240113 820916 PDR ADOCK 05000263 S PDR DDD106
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